Fissionable mass storage device

ABSTRACT

An apparatus for the safe storage of a plurality of fissionable masses including an array of discrete neutron absorbing shields which utilizes the principle of the neutron trap to reduce the multipication factor of the storage array to a subcritical value when immersed in a neutron moderating medium. Each discrete neutron absorbing shield is designed to perimetrically encircle each of the stored fissionable masses. Each shield is spaced such that the encircled fissionable mass is spaced from the next adjacent neutron absorbing shield by a distance determined by the enrichment of the fissionable masses and attenuation of the moderating medium.

REFERENCE TO RELATED APPLICATIONS

This application is a continuation-in-part of U.S. application Ser. No.558,767 filed Mar. 17, 1975 and entitled FISSIONABLE MASS STORAGE DEVICEby Frank Bevilacqua.

BACKGROUND OF THE INVENTION

The present invention relates to the safe storage of fissionable masses.More particularly the present invention relates to a seismically safearrangement for storing fissionable masses under water with a maximumstorage density without producing a critical geometry.

DESCRIPTION OF THE PRIOR ART

It is well-known to store fissionable masses such as nuclear reactorfuel element assemblies in storage pools which can accommodate eithernew fuel assemblies or spent fuel assemblies. An essential requirementfor all fissionable material storage is that the fissionable materialcannot be permitted to assume a geometry which is either critical orsupercritical. Accordingly, most if not all storage pools have devicesand mechanisms for preventing the placing of fissionable masses in suchpositions that they achieve a critical geometry. An example of such aprior art storage arrangement is disclosed in the U.S. Pat. No.3,037,120 issued to J. D. McDaniels, Jr. on May 29, 1962. However, mostprior art fuel storage devices have the disadvantages of occupying largeamounts of space in the nuclear power plant and of failing to adequatelymeet the current Nuclear Regulatory Commission seismic criteria.

FIG. 1 illustrates one such prior art nuclear reactor fuel storagearrangement. The nuclear fuel assemblies are placed in chambers 110.Chambers 110 are bounded on one pair of opposite sides by plates 112 and112' and on the other pair of opposite sides by plates 114 and 114'. Ascan be seen from the drawing, a plurality of adjacent chambers 110 sharecommon side plates 112 and 112'. Each plurality of adjacent chambers 110which share common side plates 112 and 112' are spaced from the nextplurality of adjacent chambers which also share their own common sideplates 113 and 113' by space 120. The space 120 is maintained by spacingelements 122 and 122' which are located at the extreme lateral ends ofplates 112' and 113. In a similar manner adjacent plates 114 and 114'are held apart by spacing elements 116 and 116' to establish aseparation space 118.

As above described, the piror art fissionable mass storage arrangementincludes spaces 118 and 120 which separate the storage compartments 110by a predetermined distance (d). Accordingly, if the distance (d) isproperly chosen, the walls 114, 114', 112, 112', 113, and 113' of eachchamber 110 in cooperation with the spaces 118 and 120 operate asneutron flux traps: a concept that will be discussed below. However,this prior art apparatus has the fundamental deficiency that it cannoteasily meet the seismic criteria established by the NRC. It is apparentfrom FIG. 1 tat on the occurrence of a seismic disturbance, one entiremodule, including all of the adjacent storage areas 110 which arelocated between the two steel plates 112 and 112', can move as a unit.This being the case, the unit as a whole, when filled with storedfissionable masses, could conceivably be distorted so that theseparation distance of space 120 is reduced. In such a case, storagearray may become a critical or supercritical mass since all that isrequired to obtain a critical mass is the displacement of only one ofthe fuel assemblies from its design position.

SUMMARY OF THE INVENTION

Thus is posed the problem of finding an apparatus which permits thecompact storage of fissionable masses without creating the possibilityof producing a critical geometry. The solution to the posed problemshould be such that the storage arrangement becomes simpler, lessdangerous and more compact than prior art devices. It is also desirablethat the storage arrangement be easily fabricated and easily modified tocontain fuel at different enrichments. These objects are realized by thepresent invention through a design which incorporates discrete neutronabsorbing shields adapted to perimetrically encircle each of thefissionable masses. Spacing means in at least two directions areprovided for spacing each of the discrete neutron absorbing shields fromthe next adjacent discrete nuclear absorbing shield by at least apredetermined distance which is determined by the enrichment of thefissionable mass. For the purposes of this application, this distance isdesignated the "neutron attenuation distance". The apparatus issubmerged in a storage pool under a moderator, which is ordinarilywater, and the neutron absorbing shields are open at each end so thatthe water may circulate through the interior of the shield and throughthe length of the contained fuel assembly, thereby assisting in theremoval of excess heat. The apparatus further has the feature that eachdiscrete neutron absorbing shield is flared at the upper end tofacilitate the insertion of the fuel assembly into the shield. Thepreferred embodiment of the invention includes the arrangement ofneutron absorbing shields in rows and columns such that open channelsare created between the rows and columns.

The present invention may be better understood and its numerous objectsand advantages may become apparent to those skilled in the art byreference to the accompanying drawing wherein like reference numeralsrefer to like elements in the several figures and in which

FIG. 1 is an illustration which represents the prior art;

FIG. 2 is a isometric view of the invention;

FIG. 3 is an illustration of the invention in its actual practicalapplication in a nuclear power plant storage pool; and

FIG. 4 is a pair of curves which show permissible minimum neutronattenuation distances as a function of effective enrichment of thestored fuel assemblies: curve 1 applying to storage devices made from"black" poison material and curve 2 applying to storage devices madefrom stainless steel.

DESCRIPTION OF THE PREFERRED EMBODIMENT

Recent decreases in the availability of spent fuel reprocessing plantshave created a substantial demand for increased storage facilities.Accordingly, it is desirable to design a storage facility that is ableto store a maximum number of fuel assemblies in a given volume. Whilethere exists this demand to store fuel assemblies with a maximum storagedensity, an all important and overriding requirement is that the storagearray must prevent the fissionable masses from achieving a physicalgeometry which allows the combined mass to become critical. Thisrequirement must be met in all cases and at all costs and, therefore,the spent fuel storage apparatus must prevent the creation of a criticalmass even on the occurrence of the most severe seismic disturbance. Itshould be recognized that only the inadvertent displacement of one fuelassembly is necessary to create a localized critical mass in the storagearray. The following preferred embodiment is an apparatus whichaccomplishes the object of obtaining a maximum storage density whileassuring that a critical mass is avoided in any area even on theoccurrence of a severe earthquake.

The present invention is generally illustrated in FIG. 3 which disclosesa storage array for nuclear fuel assemblies under water in a fuelstorage pool. Such storage pools are for the purpose of storing newnuclear reactor fuel or spent nuclear reactor fuel. Nuclear reactor fuelelements ordinarily have polygonal cross-sections. Spent fuel is highlyradioactive and generates considerable amounts of decay heat.Consequently, it is necessary to continuously cool the spent fuel inorder to remove the decay heat. The usual means for dissipating thedecay heat is to circulate water along the length of the fuel assemblythereby removing the heat through the mechanisms of conduction andconvection. The heated water may subsequently be removed from the spentfuel storage pool and cooled in an external heat exchanger.

The fuel storage rack of the invention as illustrated in greater detailin FIG. 2 consists of a plurality of discrete neutron absorbingcontainers 12 having polygonal cross-sections closely matching in shapeand size the polygonal cross sections of the fuel. In the preferredembodiment these containers 12 are square, open ended boxes whichclosely fit the square fuel assemblies to be stored. Each discreteneutron absorbing container 12 may be constructed by extruding a squaremetal tube of proper thickness or by welding two bent L-shaped elongatedmetal plates of proper thickness. The metal material is preferably onewhich has a relatively high neutron absorption cross-section such asstainless steel or a material having neutron absorption characteristicssimilar to the neutron absorption characteristics of a 0.63 centimeterstainless steel plate containing 1 weight percent of boron. This lattermaterial is termed a black poison or absorber and is a poison materialhaving a macroscopic absorption neutron cross-section 100 to 1000 timesgreater than stainless steel and allows the water separation necessarybetween containers to be reduced. Practical materials containing suchpoisons will require a gap which is a minimum of one-half inch for about3 weight percent enriched fuel and 1 inch for about 4 weight percentenriched fuel. Alternatively, the neutron absorbing container 12 can bemade of a material of low neutron absorption cross-section material towhich is fastened or which includes a high cross-section material suchas boron, cadmium or gadolinium. In this embodiment a square stainlesssteel container is used with a thickness ranging from 0.508 to 1.27centimeters or with a preferred thickness of 0.635 centimeters. In someapplications, the lower limit of this range can be lowered to 0.2centimeters.

In order to facilitate the deposit of fuel assemblies in these discreteclosely fitting containers, the walls of the containers are outwardlyflared at one end. This flared portion 14 is illustrated in the diagramas being the upper portion of the box. However, it should be recognizedthat while the preferred orientation of the discrete neutron absorbingcontainers is up and down, it is nevertheless possible to construct astorage rack which has its storage containers in an orientation otherthan vertical.

In the preferred embodiment a multiplicity of the discrete neutronabsorbing containers are arranged into an array of rows and columns sothat there are spaces 16, 18 both between the rows and the columns. Inthis manner, the design of the preferred embodiment includes a neutronflux trap which permits the closer spacing of fissionable masses thanwould otherwise be possible. The separating means in one direction areextended U-channels 20 to which adjacent and opposite containers arewelded. In the other direction, the adjacent containers are attached toU-channels 22 which may be abbreviated as shown in the illustration ormay be full length. These spacing means, the extended U-channels 20 andthe abbreviated U-channels 22 hold the discrete containers apart by atleast a predetermined neutron attenuation distance (a). This minimumpredetermined neutron attenuation distance (a) is a distance which isprecalculated to assure that the array of stored fuel assemblies cannotachieve a critical mass. The value of the minimum predetermined neutronattenuation distance is a function of the effective enrichment of U²³⁵in the fuel assemblies to be stored. The functional relationship appearsin FIG. 4. From FIG. 4, it can be seen that for a given enrichment andfor a black poison, values of neutron attenuation distance lying on orto the right of curve 1 results in acceptable spacing. Such values givemultiplication factors smaller than one. When the neutron absorbingcontainers 12 are stainless steel, the minimum neutron attenuationdistance is determined from curve 2 of FIG. 4.

While passing from one fuel assembly to the next adjacent fuel assemblythe neutrons must pass sequentially through a very small water gap, aplate which is one wall of the discrete neutron absorbing container 12,a gap 16 filled with a moderator such as water or borated water, and asecond plate, which is a portion of the next adjacent discrete neutronabsorbing container 12' which surrounds the next adjacent fuel assembly.It is also possible that the neutron may be reflected in gap 16 and mayreturn to the first steel plate of container 12. During its passagethrough these four mediums the typical neutron behaves as follows. Onencountering the first plate, the typical neutron is a "fast" neutronemanating from the fuel and has such a high energy that it passesthrough this first neutron absorbing material essentially unaffected andunabsorbed. During its passage through the adjacent water gap 16, thetypical neutron is moderated by the water from its high energies tolower energies to become a "slow" or low energy neutron. And finally,upon encountering the second plate or upon reencountering the firstplate after reflection in water gap 16, the neutron is absorbed, sincethe neutron has been moderated to an energy which permits the absorptionof the neutron by the neutron absorbing material. It is essential tokeep the "very small water gap" between the fuel assembly and itsenclosing container to a minimum for two reasons: first, to minimize thepossibility of the displacement of the enclosed fuel assembly from itspreferred central position; and second, to avoid the situation in whichthe neutron is moderated to low energies and is subsequently reflectedby either the water moderator or by the neutron absorbing material ofthe enclosing discrete container. It can be shown that increasing thewidth of the water gap immediately adjacent to the stored fuel assembly,increases the probability of this reflection which has the effect ofincreasing the reactivity of the array of stored fuel assemblies: anundesirable result. Accordingly, this interior gap should in no case beallowed to exceed 1.5 centimeters and the dimension of the gap 16 iscalculated by assuming the interior gap to be zero. Small interior gapsresult when the containers 12 have cross-sections closely matching inboth shape and size the cross-section of the fissionable mass to bestored.

The arrangement of adjacent fuel assemblies surrounded by discreteneutron absorbing containers creates a "neutron flux trap" in which anyneutron which is traveling from one fuel assembly through a moderator toanother fuel assembly or back to the original fuel assembly issequentially exposed to a moderating material and the trapping material.It is primarily due to this flux trap principle utilized by theinvention that the storage density of spent fuel assemblies can besubstantially increased. The dimensions involved, particularly theneutron attenuation distance through the water gap between adjacentdiscrete neutron absorbing containers, depends upon a number of factors,including the identity of the fissionable material, the fuel enrichmentof the stored fuel assembly, and the thickness of the neutron absorbingmaterial as well as the identity of the neutron absorbing material. Forthe purposes of this disclosure the predetermined neutron attenuationdistance is defined as the distance between adjacent discrete neutronabsorbing containers or the width of gap 16 as shown in FIG. 2. In thepreferred embodiment utilizing stainless steel containers the minimumpredetermined neutron attenuation distance is 9.5 centimeters for fuelassemblies having an effective enrichment of approximately 4.3 weightpercent U²³⁵. This value includes consideration of calculationuncertainty, manufacturing tolerances and a conservative criticalitymargin of 5 percent. In actual practice the prior art device illustratedin FIG. 1 and previously described ineffectively utilized the flux trapprinciple in that the separation distance built into the prior artdevice was small and the appropriate degree of neutron moderation wasnot obtained. The fuel storage apparatus of the present inventionovercomes this difficulty by providing a neutron attenuation distancewhich is sufficient to allow the water gap and the neutron absorbingshields to act effectively as a neutron flux trap. The neutronattenuation distances derivable from the curves of FIG. 4 can beincreased by as much as 8 centimeters without causing the storage arrayto depart significantly from the utilization of the neutron flux traptechnique.

The present invention includes a multiplicity of individual discreteneutron absorbing shields which individually enclose each and everystored fuel assembly. Furthermore, the device of the present inventiondiscloses an assemblage of fuel assemblies which meet the strict seismiccriteria established by the United States Nuclear Regulatory Commission.Contrary to the prior art which required the multiplicity of spent fuelassemblies which share the common walls 112 and 112' to oscillatetogether as one unit, the design of the present invention allows eachindividual stored fuel assembly to oscillate essentially independentlyof the others. Accordingly, the accumulated stresses and the permittedoscillation modes are substantially different in the present inventionfrom the prior art storage racks. As described above, the discretecontainers are spaced by U-shaped spacing members 20 and 22 at top andbottom positions. The preferred embodiment further assembles amultiplicity of storage containers into a rectangular unit 10 which hasnine containers on one side and six containers across the end. Along theoutside of this modular unit are attached angle irons 24 whichfacilitate the attachment of one modular unit to an adjacent unit bymeans of any well-known prior art device. In addition, it should berecognized that each container of the rectangular modular unit is heldabove the floor of the storage pool by a certain separation space whichallows the entry and circulation of cooling fluid through the length ofthe storage container 12. Accordingly, relatively cool coolant entersthrough the bottom of the container, flows upwardly through the lengthof the stored fuel assembly and exits through the top of the fuelassembly and the storage container 12. Furthermore, the spacing meanswhich assembles the storage containers into rows and columns permits theintermingling cross flow between the adjacent storage containers aroundthe outside thereof.

The rectangular storage units 10 containing a multiplicity of storagecontainers are assembled by setting each unit 10 on the floor of thepool or by bolting each unit 10 to a support base on the floor of thestorage pool, an bolting each adjacent rectangular unit 10 to the nextadjacent rectangular unit as described above. Depending on the seismicrestraint requirements, the rectangular units on the periphery of thestorage array are not restrained at all, are butted against the sides ofthe storage pool, or are fixedly fastened to the walls by means of angleirons 26 as illustrated in FIG. 3. In the alternative, the separatingU-channels may extend outwardly of the rectangular unit permitting theexternal attachment to the side spacing means. Also as an alternative oran additional feature that increases seismic resistance, additionalspacing means may be provided between the adjacent containers midway ofthe length of the containers and/or diagonal spacers which diagonallytraverse a number of containers in the same row or column. By way ofspecific example, storage of nuclear reactor fuel assemblies having anenrichment of 4.3 weight percent of U²³⁵ and having an outside diameterof 22.86 centimeters may be stored in a device of the present inventionwhose stainless steel neutron absorbing shield thickness is 0.635centimeters and whose minimum neutron attenuation distance is 9.46centimeters. For nuclear reactor fuel assemblies having an enrichment of3.5 weight percent of U²³⁵ and having an outside diameter of 24.99centimeters, an appropriate storage device would be a device asdescribed hereinbefore whose stainless steel neutron absorbing shieldthickness is 0.635 centimeters and whose minimum neutron attenuationdistance is 8.81 centimeters.

What is claimed is:
 1. An apparatus for the safe yet compact storage ofa plurality of fissionable masses having a predetermined effectivefissile enrichment and having cross-sections of a given size and shape,comprising:a. a plurality of adjacent discrete neutron absorbing openended shields disposed in side-by-side relationship, one of each ofwhich is adapted to parametrically encircle one of each of saidplurality of fissionable masses, said neutron absorbing shields havingneutron absorbing characteristics similar to the neutron absorbingcharacteristics of a 0.635 centimeter thick stainless steel platecontaining one weight percent of natural boron and having cross-sectionsclosely matching in size and shape the cross-sections of the fissionablemasses to be stored; b. spacing means in each of two directions forspacing each of said side-by-side neutron absorbing shields from thenext adjacent neutron absorbing shield by a predetermined minimumneutron attenuation distance no less than the minimum neutronattenuation distance determined from curve 1 of FIG. 4 for saidpredetermined effective fissile enrichment; and c. hydrogenousmoderating means between said adjacent discrete neutron absorbingshields for moderating neutrons which are emitted from said fissionablemasses whereby a neutron flux trap is created.
 2. The apparatus asrecited in claim 1 wherein said moderating means is water.
 3. Theapparatus as recited in claim 1 wherein said spacing means includes atleast two spaced apart spacing means for each side of each of saidneutron absorbing shields, one pair of spacing means being near one endof said neutron absorbing shield and the other pair of said spacingmeans being near the other end of the said neutron absorbing shield. 4.An apparatus for the safe yet compact storage of a plurality offissionable masses having a predetermined effective fissile enrichmentand having cross-sections of a given size and shape, comprising:a. aplurality of adjacent discrete neutron absorbing open ended shieldsdisposed in side-by-side relationship, one of each of which is adaptedto parametrically encircle one of each of said plurality of fissionablemasses, said neutron absorbing shields consisting of stainless steel andhaving cross-sections closely matching in size and shape thecross-section of the fissionable masses to be stored; b. spacing meansin each of two directions for spacing each of said side-by-side neutronabsorbing shields from the next adjacent neutron absorbing shield by apredetermined minimum neutron attenuation distance no less than theminimum neutron attenuation distance determined from curve 2 of FIG. 4for said predetermined effective fissile enrichment; and c. hydrogenousmoderating means between said adjacent discrete neutron absorbingshields for moderating neutrons which are emitted from said fissionablemasses whereby a neutron flux trap is created.
 5. The apparatus asrecited in claim 4 wherein said moderating means is water.
 6. Theapparatus as recited in claim 4 wherein said spacing means includes atleast two spaced apart spacing means for each side of each of saidneutron absorbing shields, one pair of spacing means being near one endof said neutron absorbing shields and the other pair of said spacingmeans being near the other end of said neutron absorbing shields.
 7. Theapparatus as recited in claim 4 wherein said stainless steel of saiddiscrete neutron absorbing shields has a thickness in the range from 0.2centimeters to 1.4 centimeters.